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Heat treatments of irradiated uranium oxide in a pressurised water reactor (P.W.R.): swelling and fission gas release; Traitements thermiques de l`oxyde d`uranium irradie dans un reacteur a eau pressurisee (R.E.P.): gonflement et relachement des gaz de fission

Abstract

In order to keep pressurised water reactors at a top level of safety, it is necessary to understand the chemical and mechanical interaction between the cladding and the fuel pellet due to a temperature increase during a rapid change in reactor. In this process, the swelling of uranium oxide plays an important role. It comes from a bubble precipitation of fission gases which are released when they are in contact with the outside. Therefore, the aim of this thesis consists in acquiring a better understanding of the mechanisms which come into play. Uranium oxide samples, from a two cycles irradiated fuel, first have been thermal treated between 1000 deg C and 1700 deg C for 5 minutes to ten hours. The gas release amount related to time has been measured for each treatment. The comparison of the experimental results with a numerical model has proved satisfactory: it seems that the gases release, after the formation of intergranular tunnels, is controlled by the diffusion phenomena. Afterwards, the swelling was measured on the samples. The microscopic examination shows that the bubbles are located in the grain boundaries and have a lenticular shape. The swelling can be explained by the bubbles coalescence and  More>>
Authors:
Publication Date:
Mar 27, 1997
Product Type:
Technical Report
Report Number:
CEA-R-5769
Reference Number:
SCA: 210200; PA: AIX-29:065651; EDB-99:008464; SN: 99002049336
Resource Relation:
Other Information: DN: 56 refs.; PBD: 27 Mar 1997
Subject:
21 NUCLEAR POWER REACTORS AND ASSOCIATED PLANTS; BENCH-SCALE EXPERIMENTS; CREEP; DATA ACQUISITION; FISSION PRODUCT RELEASE; FUEL CANS; FUEL PELLETS; GRAIN BOUNDARIES; HEAT TREATMENTS; IMAGE PROCESSING; POROSITY; PWR TYPE REACTORS; RADIATION PROTECTION; SCANNING ELECTRON MICROSCOPY; STRESS CORROSION; SWELLING; THERMAL EXPANSION; URANIUM OXIDES
OSTI ID:
296079
Research Organizations:
CEA Centre d`Etudes de Saclay, 91 - Gif-sur-Yvette (France). Dept. de Mecanique et de Technologie; Ecole Centrale de Paris, 75 (France)
Country of Origin:
France
Language:
French
Other Identifying Numbers:
Other: ON: DE99609720; TRN: FR9803351065651
Availability:
INIS; OSTI as DE99609720
Submitting Site:
FRN
Size:
230 p.
Announcement Date:

Citation Formats

Zacharie, I. Heat treatments of irradiated uranium oxide in a pressurised water reactor (P.W.R.): swelling and fission gas release; Traitements thermiques de l`oxyde d`uranium irradie dans un reacteur a eau pressurisee (R.E.P.): gonflement et relachement des gaz de fission. France: N. p., 1997. Web.
Zacharie, I. Heat treatments of irradiated uranium oxide in a pressurised water reactor (P.W.R.): swelling and fission gas release; Traitements thermiques de l`oxyde d`uranium irradie dans un reacteur a eau pressurisee (R.E.P.): gonflement et relachement des gaz de fission. France.
Zacharie, I. 1997. "Heat treatments of irradiated uranium oxide in a pressurised water reactor (P.W.R.): swelling and fission gas release; Traitements thermiques de l`oxyde d`uranium irradie dans un reacteur a eau pressurisee (R.E.P.): gonflement et relachement des gaz de fission." France.
@misc{etde_296079,
title = {Heat treatments of irradiated uranium oxide in a pressurised water reactor (P.W.R.): swelling and fission gas release; Traitements thermiques de l`oxyde d`uranium irradie dans un reacteur a eau pressurisee (R.E.P.): gonflement et relachement des gaz de fission}
author = {Zacharie, I}
abstractNote = {In order to keep pressurised water reactors at a top level of safety, it is necessary to understand the chemical and mechanical interaction between the cladding and the fuel pellet due to a temperature increase during a rapid change in reactor. In this process, the swelling of uranium oxide plays an important role. It comes from a bubble precipitation of fission gases which are released when they are in contact with the outside. Therefore, the aim of this thesis consists in acquiring a better understanding of the mechanisms which come into play. Uranium oxide samples, from a two cycles irradiated fuel, first have been thermal treated between 1000 deg C and 1700 deg C for 5 minutes to ten hours. The gas release amount related to time has been measured for each treatment. The comparison of the experimental results with a numerical model has proved satisfactory: it seems that the gases release, after the formation of intergranular tunnels, is controlled by the diffusion phenomena. Afterwards, the swelling was measured on the samples. The microscopic examination shows that the bubbles are located in the grain boundaries and have a lenticular shape. The swelling can be explained by the bubbles coalescence and a model was developed based on this observation. An equation allows to calculate the intergranular swelling in function of time and temperature. The study gives the opportunity to predict the fission gases behaviour during a fuel temperature increase. (author) 56 refs.}
place = {France}
year = {1997}
month = {Mar}
}